ML17170A183

Issuance of Amendments Regarding the Extension of the Emergency Service Water Pump Allowed Outage Time, Surry Power Station, Units 1 and 2

ML17170A183
Person / Time
Site:	Surry
Issue date:	07/28/2017
From:	Cotton K Plant Licensing Branch 1
To:	Stoddard D Virginia Electric & Power Co (VEPCO)
Cotton K R/415-1438
References
CAC MF8145, CAC MF8146
Download: ML17170A183 (24)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 28, 2017 Mr. Daniel G. Stoddard President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Blvd.

Glenn Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING THE EXTENSION OF THE EMERGENCY SERVICE WATER PUMP ALLOWED OUTAGE TIME, SURRY POWER STATION, UNITS 1 AND 2 (CAC NOS. MF8145 AND MF8146)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 290 to Renewed Facility Operating License No. DPR-32 and Amendment No. 290 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively.

The amendments revise the Technical Specifications (TSs) in response to your application dated July 14, 2016, and as supplemented by letters dated January 31, 2017, March 1, 2017, and March 10, 2017.

These amendments revise TS 3.14, "Circulating and Service Water Systems," to extend the Allowed Outage Time (AOT) for Emergency Service Water (ESW) pump inoperability. The proposed revision extends the TS 3.14.B AOT for one inoperable ESW pump from 7 to 14 days to provide operational flexibility for ESW pump maintenance and repairs.

D. Stoddard A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Karen Cotton Gross, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 290 to DPR-32

2. Amendment No. 290 to DPR-37

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 290 Renewed License No. DPR-32

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated July 14, 2016, as supplemented by letter dated January 31, 2017, March 1, 2017, and March 10, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: July 28, 201 7.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 290 Renewed License No. DPR-37

1. The Nuclear Regulatory Commission (the Commission) has found that:

A The application for amendment by Virginia Electric and Power Company (the licensee) dated July 14, 2016, as supplemented by letter dated January 31, 2017, March 1, 2017, and March 10, 2017, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3. B of Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance July 28, 2017.

ATTACHMENT TO LICENSE AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE DPR-32 LICENSE AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE DPR-37 SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

- Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-37, page 3 License No. DPR-37, page 3 TSs TSs TS 3.14-2 TS 3.14-2 TS 3.14-3 TS 3.14-3

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 I. Fire Protection The licensee shall implement and maintain in effect the provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report and as approved in the SER dated September 19, 1979, (and Supplements dated May 29, 1980, October 9, 1980, December 18, 1980, February 13, 1981, December 4, 1981, April 27, 1982, November 18, 1982, January 17, 1984, February 25, 1988, and Surry - Unit 1 Renewed License No. DPR-32 Amendment No. 290

E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such by product and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

J. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power Levels not in excess of 2587 megawatts (thermal)

K. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290 are hereby incorporated in this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

L. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

M. Records The licensee shall keep facility operating records in accordance with the Requirements of the Technical Specifications.

N. Deleted by Amendment 54

0. Deleted by Amendment 59 and Amendment 65 P. Deleted by Amendment 227 Q. Deleted by Amendment 227 Surry - Unit 2 Renewed License No. DPR-37 Amendment No. 290

TS 3.14-2

5. Two service water flow paths to the charging pump service water subsystem are OPERABLE.

6. Two service water flow paths to the recirculation spray subsystems are OPERABLE.

7. Two service water flow paths to the main control room and emergency switchgear room air conditioning subsystems are OPERABLE.

B. The requirements of Specification 3.14.A.4 may be modified to allow one Emergency Service Water pump to remain inoperable for a period not to exceed 14 days. If this pump is not OPERABLE in 14 days, then place both units in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The requirements of 3 .14.A.4 may be mod]fied to have two Emergency Service Water pumps OPERABLE with one unit in COLD SHUTDOWN with combined Spent Fuel pit and shutdown unit decay heat loads of 25 million BTU/HR or less.

One of the two remaining pumps may be inoperable for a period not to exceed 14 days. If this pump is not OPERABLE in 14 days,. then place the operating unit in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

C. The requirements of Specifications 3.14.A.5 and 3.14.A.7 may be modified to allow unit operation with only one OPERABLE flow path to the charging pump service water subsystem and to the main control and emergency switchgear rooms air conditioning condensers. If the affected systems are not restored to the requirements of Specifications 3.14.A.5 and 3.14.A.7 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the requirements of Specifications 3.14.A.5 and 3.14.A.7 are not satisfied as allowed by this Specification, the reactor shall be placed in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

D. The requirements of Specification 3.14.A.6 may be modified to allow unit operation with only one OPERABLE flow path to the recirculation spray subsystems. If the affected system is not restored to the requirements of Specification 3.14.A.6 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the requirements of Specification 3.14.A.6 are not met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Amendment Nos. 290 and 290

TS 3.14-3 Basis The Circulating and Service Water Systems are designed for the removal of heat resulting from the operation of various systems and components of either or both of the units.

Untreated water, supplied from the James River and stored in the high level intake canal is circulated by gravity through the recirculation spray coolers and the bearing cooling water heat exchangers and to the charging pumps lubricating oil cooler service water pumps which supply service water to the charging pump lube oil coolers.

In addition, the Circulating and Service Water Systems supply cooling water to the component cooling water heat exchangers and to the main control and emergency switchgear rooms air conditioning condensers. The Component Cooling heat exchangers are used during normal plant operations to cool various station components and when in shutdown to remove residual heat from the reactor. Component Cooling is not required on the accident unit during a loss-of-coolant accident. If the loss-of-coolant accident is coincident with a loss of off-site power, the nonaccident unit will be maintained at HOT SHUTDOWN with the ability to reach COLD SHUTDOWN.

The long term Service Water requirement for a loss-of-coolant accident in one unit with simultaneous loss-of-station power and the second unit being brought to HOT SHUTDOWN is greater than 15,000 gpm. Additional Service Water ls necessary to bring the nonaccident unit to COLD SHUTDOWN. Three diesel driven Emergency Service Water pumps with a design capacity of 15,000 gpm each, are provided to supply water to the High Level Intake canal during a loss-of-station power incident. Thus, considering the single active failure of one pump, three Emergency Service Water pumps are required to be OPERABLE. The allowed outage time of 14 days provides operational flexibility to allow for repairs up to and

Amendment Nos. 290 and 290

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1AND2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated July 14, 2016, and as supplemented by letters dated January 31, 2017, March 1, 2017, and March 10, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML16202A068, ML17037D053, ML17066A187, and ML17075A256, respectively), Virginia Electric and Power Company (the licensee) submitted a request for changes to the Surry Power Station, Unit Nos. 1 and 2 (Surry) Technical Specifications (TSs).

The requested changes would extend the Allowed Outage Time (AOT) for Emergency Service Water (ESW) pump inoperability. The supplements dated January 31, 2017, March 1, 2017, and March 10, 2017, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination that was published in the Federal Register (FR) on October 25, 2016 (81 FR 73443).

The proposed changes would revise TS 3.14, "Circulating and Service Water Systems," to extend the AOT for ESW pump inoperability. The proposed revision extends the TS 3.14.B AOT for one inoperable ESW pump from 7 to 14 days to provide operational flexibility for ESW pump maintenance and repairs. The proposed revision extends the TS 3.14. B AOT for one inoperable ESW pump from 7 to 14 days for the specified plant operating conditions (i.e., both units operating or one operating and one in cold shutdown with specified heat loads).

2.0 REGULATORY EVALUATION

2.1 System Description The Surry service water system (SWS) is described in Section 9.9 of the Updated Final Safety Analysis Report. The SWS function is supported by two pumping systems at Surry: the plant circulating water (CW) pumps and the ESW pumps. During normal operation, the CW pumps pump water from the James River up to the high level intake canal. Four motor-driven CW pumps for each unit are located in the low level intake structure. When power is not available to the CW pumps, three diesel-driven ESW pumps located in the low level intake structure ensure an adequate water supply to the high level intake canal to support the SWS functions.

Enclosure 3

Each ESW pump assembly consists of a diesel engine connected to a vertical turbine-type pump through an angled reduction gear drive. Each pump is located in a separate CW pump screenwell and is operated locally. Each diesel engine is started by a battery-powered electric motor actuated by a local pushbutton. The diesel fuel supply is a shared tank with adequate capacity for operation of three ESW pumps for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and two pumps for an additional 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The high-level canal is part of the flow path for the CW system and SWS, and acts as a reservoir for these systems. The high-level canal is about 1. 7 miles long and is concrete-lined along its entire length. The canal sides slope from top to bottom with an average width of 125 feet, and the canal bottom slopes from an elevation of 6 feet 8 inches at the low level intake end, by the river, to an elevation of 5 feet just outside the high level intakes for the two units.

The canal floor slopes rapidly from 5 feet to an elevation of 8 inches at each unit's intake. The canal is normally maintained at a water surface elevation of approximately 25 to 30 feet.

TS 3.14.A.1 requires a minimum elevation of 23 feet when the reactor is critical or the reactor coolant system exceeds a temperature of 350 degrees Fahrenheit or a pressure of 450 pounds per square inch gauge.

Water flows by gravity from the high level intake canal through each unit's high level intake and the unit heat exchangers to the discharge canal that returns water to the James River. The CW flows by gravity from the high-level intake canal through four buried parallel lines to each condenser and then through four separate lines to the discharge tunnel for each unit. The SWS piping branches from the CW piping upstream of the main condenser and delivers water to the various heat exchangers and services cooled by the SWS. The CW and SWS provide cooling water to the following equipment: main condenser waterboxes (4 per unit with a motor-operated valve (MOV) at the inlet and outlet of each waterbox); bearing cooler heat exchangers (3 per unit with an MOV in each of two supply lines); recirculation spray heat exchangers (4 per unit with 2 parallel MOVs in each of two supply lines to the heat exchanger supply header and an MOV at both the inlet and outlet of each heat exchanger); and component cooling heat exchangers (4 shared heat exchangers with an MOV in each of two supply lines).

In addition, three branch lines from the CW piping provide water to the charging pump SWS (1 per unit) and the shared booster pumps that provide cooling to the control room and electric relay room air conditioning condensers. Water from this equipment empties into one of the unit discharge tunnels, which then flows into the discharge canal.

The safety function of the SWS is to remove residual heat and heat rejected by certain safety-related equipment. The recirculation spray heat exchangers remove residual heat from primary containment following a design basis accident (OBA) that releases mass and energy into the primary containment. The component cooling water heat exchangers provide for shutdown cooling of the reactors and decay heat removal from the spent fuel pool. The charging pump SWS cools the lube oil and bearings for the charging pumps that are used for primary makeup and high-head safety injection. The air conditioning condensers reject heat from the control room and electrical relay rooms.

The recirculation spray heat exchanger MOVs open on a high-high containment pressure signal indicative of a loss-of-coolant accident (LOCA). The parallel valves in the supply lines to the recirculation spray heat exchanger supply header ensure that a single failure of a valve would result in the loss of service water flow to no more than one recirculation spray heat exchanger.

The water inventory available in the high-level intake canal and additional flow from the ESW pumps provide an assured source of cooling water. A canal elevation of 17 .2 feet is the minimum elevation necessary to provide design flow of service water through the recirculation spray heat exchangers during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA.

Automatic isolation of non-essential service water flow paths ensures an adequate inventory of water in the intake canal for essential functions. In the event of a LOCA coincident with a loss of off-site power (LOOP), the supply to the recirculation spray heat exchangers for the affected unit will open, and all other service water isolation valves will close for the affected unit. If the canal water level falls below 23.5 feet, signals are generated to trip both units' turbines, and to close the nonessential condenser CW MOVs and the service water supply MOVs for the bearing cooler heat exchangers, and the component cooling heat exchangers.

A failure of an MOV in the supply lines to either the component cooling heat exchanger or the bearing cooler heat exchangers would increase the rate of water loss, but redundant manual valves are located in accessible areas to permit local isolation of service water flow to these components. When shutdown cooling of a reactor or cooling of the spent fuel pool is necessary, operating procedures provide for manual throttling of the service water supply valves to the component cooling heat exchangers. In addition, passive vacuum breakers are installed on the CW pump discharge lines to assure that a reverse siphon is not continued for canal levels less than 23 feet when the CW pumps are de-energized. The remaining six feet of canal level plus water added by ESW pump operation provides the required source of service water following a OBA.

Regarding ESW pump operation following a OBA or loss of power, the TS Bases for TS 3.14.A.4 state that the long term Service Water requirement for a LOCA accident in one unit with simultaneous loss-of-station power and the second unit being brought to hot shutdown is greater than 15,000 gallons per minute (gpm). Additional Service Water is necessary to bring the non-accident unit to cold shutdown. Three diesel driven ESW pumps with a design capacity of 15,000 gpm each, are provided to supply water to the High Level Intake canal during a loss-of-station power incident. Thus, considering the single active failure of one pump, three ESW pumps are required to be operable. When one of the two units has been cooled down, and the total heat load from that unit and the spent fuel pit drops below 25 million British thermal units per hour (BTU/hr), a single ESW pump can provide the long-term service water requirements for the facility. Accordingly, under these conditions, TS 3.14.B requires just two ESW pumps to be operable.

The required start time of the ESW pumps is dependent on the heat exchangers requiring service water flow and any failures of automatic isolation that allow continued service water flow to unnecessary heat exchangers. In Section 5, "Probabilistic Risk Assessment," of the licensee's application dated July 14, 2016, the licensee stated that, for both a OBA (i.e., a large break LOCA with coincident LOOP) and a dual unit LOOP or other condition causing a loss of all CW pumps, the retained inventory in the intake canal would provide adequate service water flow for the 24-hour mission time modeled in the risk assessment. This timing was based on the following conditions: service water flow to the component cooling heat exchangers is successfully throttled in accordance with operating procedures to manage intake canal level; and service water flow to the bearing cooling heat exchangers and main condenser waterboxes is successfully isolated by the automatic isolation valves. If either condition is not satisfied, the ability to supply service water over the 24-hour mission time from the existing canal inventory would be challenged, and operation of one or more ESW pumps before the end of the 24-hour period would be necessary to supply water to the intake canal.

The safety significance of the ESW pumps at Surry is lower than that of comparable cooling water pumps at other large light-water reactors. Unlike many other facilities, the emergency onsite alternating current power sources (i.e., diesel generators) are not dependent on the SWS for cooling at Surry. In addition, the large reservoir of water in the intake canal allows substantial time to start an ESW pump or take other actions to maintain adequate level for the SWS functions.

2.2 Proposed Change In its application dated July 14, 2016, the licensee proposed to modify TS 3.14.B to extend the AOT for an inoperable ESW pump from 7 days to 14 days, as follows, with proposed changes noted in bold bracketed text:

The requirements of Specifications 3.14.A.4 may be modified to allow one Emergency Service Water pump to remain inoperable for a period not to exceed 7 days [being revised to 14 days]. If this pump is not OPERABLE in 7 days

[being revised to 14 days], then place both units in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The requirements of 3.14.A.4 may be modified to have two Emergency Service Water pumps OPERABLE with one unit in COLD SHUTDOWN with combined Spent Fuel pit and shutdown unit decay heat loads of 25 million BTU/HR or less.

One of the two remaining pumps may be inoperable for a period not to exceed 7 days [being revised to 14 days]. If this pump in not OPERABLE in 7 days

[being revised to 14 days], then place the operating unit in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The proposed change also included corresponding changes to the TS 3.14 Bases.

2.3 Applicable Regulatory Criteria Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, "Technical specifications,"

requires that each operating license issued by the Commission contain TSs that include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met. The remedial action for an inoperable component or system has an associated completion time, which is referred to as the AOT at Surry.

The NRC staff used the following regulatory requirements and guidance to review the risk-informed AOT extension request.

• 10 CFR 50.36, Paragraph (c)(2)(ii)(C), Criterion 3, requires that LCOs be established for a structure, system, or component (SSC) that is part of the primary success path and which functions or actuates to mitigate the OBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

• Standard Review Plan (SRP), Chapter 16.1, "Risk-Informed Decision Making: Technical Specifications," contains five key principles of the NRC staff's philosophy of risk-informed decision making. They are: (1) the proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change; (2) the proposed change is

consistent with the defense-in-depth philosophy; (3) the proposed change maintains sufficient safety margins; (4) when proposed changes result in an increase in core damage frequency or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement; and (5) the impact of the proposed change should be monitored using performance measurement strategies.

• SRP, Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant Specific Changes to the Licensing Basis: General Guidance," provides general guidance for evaluating the technical basis for proposed risk-informed changes. Guidance on evaluating probabilistic risk assessment (PRA) technical adequacy is provided in SRP, Chapter 19, Section 19.1, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests after Initial Fuel Load."

• RG 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," (ADAMS Accession No. ML100910006), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed permanent licensing-basis changes by considering engineering issues and applying risk insights. This regulatory guide also provides risk acceptance guidelines for evaluating the results of such evaluations.

• RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," March 2009, (ADAMS Accession No. ML090410014) describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decisionmaking for light-water reactors.

The NRC staff also performs its review of risk-informed changes to TS requirements in accordance with the guidance provided by SRP Chapter 16.1. SRP Chapter 16.1 refers to Regulatory Guide (RG) 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," May 2011 (ADAMS Accession No. ML100910008), as an acceptable approach for assessing proposed risk-informed changes to TS AOTs.

One acceptable approach for making risk-informed decisions about proposed TS changes, including both permanent and temporary TS changes, is to show that the proposed changes meet the following five key principles listed in RG 1.177:

1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change.

2. The proposed change is consistent with the defense-in-depth philosophy.

3. The proposed change maintains sufficient safety margins.

4. When the proposed changes result in an increase in core-damage frequency (CDF) or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

5. The impact of the proposed changes should be monitored using performance measurement strategies.

The first three principles pertain to traditional engineering considerations, and the last two principles involve risk considerations.

3.0 TECHNICAL EVALUATION

3.1. Traditional Engineering Considerations 3.1.1 Key Principle 1: Compliance with Current Regulations The regulation pertinent to the licensee's proposed TS amendment request is 10 CFR 50.36(c)(2)(ii)(C), Criterion 3. The licensee proposed a permanent change to TS 3.14.B to increase the AOT from 7 days to 14 days for one inoperable ESW pump The requested change does not propose any deviation or exemption to the regulation itself but rather a permanent change to how the regulation is implemented. Therefore, the staff concludes that the proposed change is in compliance with current regulations.

3.1.2 Key Principle 2: Defense-In-Depth The staff reviewed the information contained in the licensee's application and its supplements, and evaluated the information against the defense-in-depth attributes included in RG 1.177.

The staff determined the attributes were adequately satisfied by evaluating the change to the following criteria in RG 1.777.

A reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.

The primary accident safety functions of the ESW system are to provide cooling for safety-related equipment relied upon to prevent core damage and to remove heat to prevent containment failure. The proposed change does not affect the balance between prevention of core damage and prevention of containment failure because the system continues to support all design functions. Although the reliability of the functions is reduced by a decrease in redundancy during the proposed extension to the time allowed to restore an inoperable ESW pump to operable status, a reasonable balance among these performance goals would be maintained by the inventory of water maintained in the intake canal and the remaining operable ESW pump(s).

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

The proposed change does not change the design and operation of the SWS or the ESW pumps. Therefore, the staff concludes that the proposed TS change would not involve an overreliance on programmatic activities and is acceptable.

System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers).

The proposed change increases the length of time redundancy is reduced. However, the proposed increase in the AOT to restore one inoperable ESW pump to operable status is explicitly modeled in the probabilistic risk assessment. Because the risk assessment shows changes in safety metrics fall within acceptable bounds, it provides reasonable assurance no risk outliers are introduced due to inadequate system redundancy, independence, or diversity.

Defense against potential common cause failures is preserved, and the potential for the introduction of new common cause failure mechanisms is assessed.

Neither the mode of operation, nor the configuration of the SWS is changed by the proposed TS changes. The system will continue to be operated, maintained, and tested in the same manner as before and is acceptable.

The increase in the AOT for one ESW pump from 7 days to 14 days may change the scope of corrective maintenance performed during plant operation. The staff considered the potential for corrective maintenance to involve precursors to ESW pump failure and the possibility that the underlying condition could affect more than one ESW pump. In the licensee's response to a request for additional information provided by letter dated January 31, 2017, Dominion listed the following strategies implemented at Surry to ensure common cause failure potential of the ESW pumps is minimized:

• periodic inservice testing of the ESW pumps,

• routine preventive maintenance items,

• daily operator rounds,

• diver inspections and pump condition monitoring during periods of greatest potential for biological fouling, and

• corrective action measures to evaluate extent of condition.

The staff concludes that these strategies are appropriate for early identification of conditions that could affect pump reliability and promote prompt corrective action before the conditions could lead to a common cause failure (i.e., a failure affecting both the pump undergoing maintenance as well as an on-demand failure of a pump considered operable). Therefore, the staff concludes that the proposed TS change and associated maintenance activities would neither degrade existing protections against common mode failures nor introduce credible new common cause failure mechanisms and, therefore, are acceptable.

Independence of barriers is not degraded.

The relationship of the SWS to individual fission product barriers, such as the fuel cladding, the reactor coolant system pressure boundary, and containment, will not change as a result of the proposed TS change. Therefore, the proposed TS change would not degrade the independence of these barriers and is acceptable.

Defense against human errors is preserved.

Operator response is not expected to change during normal, abnormal or emergency operating conditions. Therefore, the staff finds that the proposed TS change satisfies this attribute and is acceptable.

The intent of the General Design Criteria in Appendix A to 10 CFR, Part 50, is maintained.

The proposed change does not involve any physical changes to the design of the SWS or the ESW pumps. Therefore, the NRC staff concludes that the proposed change satisfies this attribute and is acceptable.

3.1.3 Key Principal 3: Safety Margins The proposed change does not modify or otherwise impact codes and standards that are applicable to the SWS. The SWS is not being physically modified, and the proposed AOT extension does not result in an unreasonable decrease in the availability of a redundant train of the SWS. Therefore, an adequate margin of safety will be maintained.

3.2 Risk Considerations 3.2.1 Key Principle 4: Change in Risk Consistent with the Commission's Safety Goal Policy Statement The following evaluation addresses the NRC staff's philosophy of risk-informed decision making. For proposed changes resulting in a change in CDF or risk, the increase should be small and consistent with the intent of the Commission's Safety Goal Policy Statement. The licensee stated that the Surry PRA was used to evaluate the impact of the change on CDF and Large Early Release Frequency (LERF) to support requesting an extension to the AOT to allow only one operable ESW pump from 7 days to 14 days. The ESW pumps are safety-related and shared across both Surry units and, therefore, are explicitly modeled in the average maintenance model for Surry.

PRA Technical Adequacy RG 1.174, Revision 2 states that the scope, level of detail, and technical adequacy of the PRA are to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision-making process.

The acceptability of the PRA must be compatible with the safety implications of the requested TS change and the role that the PRA plays in justifying that change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor must go into ensuring the acceptability of the PRA. This applies to Tier 1 as well as to Tiers 2 and 3 to the extent that a PRA model is used.

RG 1.200, Revision 2 describes one acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results such that the PRA can be used in regulatory decision-making for light-water reactors. RG 1.200, Revision 2 clarifies the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard to be ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications."

Internal Events PRA (Includes Internal Flooding)

The licensee reported in its submittal dated July 14, 2016, that its PRA underwent a 1998 Nuclear Energy Institute PRA peer review, a 2009 Surry PRA self-assessment, a 2010 PRA focused scope peer review, and a 2012 PRA focused scope peer review. In its letter dated March 1, 2017, the licensee responded to NRC RAls 4 and 5 and stated that the 1998 review was completed using the Westinghouse Owners Group (WOG) Peer Review Process (i.e., prior to the issuance of RG 1.200) and that the 2009 self-assessment was performed by reviewing the Surry internal events PRA model files and documentation against the requirements in the PRA standard RA Sb 2005 and RG 1.200 Revision 1. RG 1.200, Revision 2 discusses the NRC

expectation that if the results of a self-assessment are used to demonstrate the technical adequacy of a PRA for an application, differences between the current version of the standard (i.e., RA- Sa -2009) as endorsed in RG 1.200 Revision 2, Appendix A and the earlier version be identified and addressed.

In Table RAI Sb of its letter dated March 1, 2017, the licensee responded to NRC RAI 5 and identified supporting requirements that differed between RA-Sb-2005 and RA-Sb-2009 and reevaluated all the original Facts and Observations (F&Os) against the RA-Sb-2009 version of the PRA Standard. The 2010 and 2012 focused scope peer reviews were performed against the requirements in the PRA standard RA-Sa-2009 and RG 1.200, Revision 2. The NRC staff concludes that the licensee has reviewed its PRA consistent with RG 1.200 because the 2010 and 2012 peer reviews used the current PRA standard and the earlier peer review results were updated to be consistent with the current PRA standard.

In its letter dated March 10, 2017, the licensee responded to RAI 7 and provided additional information encompassing all of the findings and dispositions generated from the Surry PRA peer reviews. In addition, in its responses to RAls 4a and 5, the licensee provided information pertaining to the version of the standard against which the peer reviews and independent assessment were performed and additional details to support the scope of the peer reviews (i.e., Technical Elements and High Level Requirements). The NRC staff concludes that the licensee's PRA peer review scope, disposition of the F&Os, and responses to the NRC RAls provide sufficient confidence that the PRA is acceptable to support the risk analysis for the permanent extension of the AOT for TS 3.14. B. The NRC staff concludes that the licensee has reviewed its PRA consistent with the guidance in RG 1.200, Revision 2.

In its letter dated March 1, 2017, the licensee responded to NRC RAI 1 and provided supplemental information on the impact and the resolution of the peer reviewed F&Os addressed by three sensitivities. The sensitivity analyses involved addressing findings pertaining to (1) human error probability (HEP) values used in the model; (2) common cause factor (CCF) probabilities; and (3) LOOP frequencies. The licensee stated that the sensitivity analyses multiplied the associated probability and frequency values by a factor of ten. The results of the sensitivity analysis for core damage frequency (~CDF), large early release frequency (~LERF), incremental conditional core damage probability (ICCDP), and incremental conditional large early release probability (ICLERP)) provided in Table 1 of its RAI response remain well below the acceptance criteria in RG 1.174, Revision 2 and RG 1.177, Revision 1, and provide confidence that the bounding analyses associated with the probability values would not change the conclusions for this requested extension of the ESW pump AOT.

The change in risk to extend the ESW pump AOT is very small. The licensee's sensitivity study addressing unresolved findings for the ESW pump AOT demonstrated that the very small change in risk reported for the AOT extension is increased given conservative changes, but the results remain well below the acceptance guidelines. The staff has also reviewed the discussion of unresolved findings in Table 7 of the licensee's RAI response dated March 1, 2017, and determined that the licensee's conclusion of minimal impact to the ESW pump AOT was satisfactory for the associated change. Therefore the NRC staff concludes that the licensee's internal events PRA is acceptable in accordance with RG 1.200, Revision 2 and RG 1.174, Revision 1.

Fire PRA The licensee reviewed its Individual Plant Examination for External Events (IPEEE) and Fire Contingency Action (FCA) procedures to evaluate the impact of the extension of the AOT with only one ESW pump operable for 14 days. The NRC reviewed the Surry IPEEE and reported its review results in a memorandum from T. King to J. Zwolinski, NRC, titled, "Review of Surry Power Station Units 1 & 2 Individual Plant Examination of External Events (IPEEE) Submittal,"

dated March 7, 2000 (ADAMS at Accession No. ML17142A410). In this memorandum, the NRC concluded that on the basis of the IPEEE review, the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities.

Section 3.1 of RG 1.200, Revision 2 states that missing hazard groups may be evaluated using bounding arguments to cover the risk contributions not addressed by the model. The licensee evaluated each of the non-screened fire areas. The licensee states that in review of the IPEEE significant core damage sequences, the ESW pumps do not perform a function for these sequences. The CW system provides cooling water to the station, and the ESW pumps ensure that cooling water can be provided when power to the CW pumps is not available. In review of the Appendix R report the licensee states that ESW pumps and CW pumps are considered to be in two separate areas, therefore a fire that would render the ESW pumps inoperable would not affect the CW pumps and vice versa.

In its letter dated March 1, 2017, the licensee responded to NRC RAI 8 and provided additional quantitative analysis to address the staff's request for a bounding estimate that considered fire-induced failures of equipment, cables, and spurious operation for the identified event scenarios. The licensee used the total frequency provided in NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology" (this document is not publically available), and divided it across the four scenarios to determine a bounding frequency value for each scenario. The licensee also stated that the determined fire risk using the frequency values did not credit fire suppression or the current plant configuration that includes reactor coolant pump low leakage seals. The results provided in Table RAl8a2 for the bounding analysis in the licensee's letter dated March 1, 2017, meets the acceptance criteria for LiCDF, LiLERF, ICCDP, and ICLERP in RG 1.174, Revision 1, and RG 1.200, Revision 2.

The NRC staff concludes that the qualitative assessment of the significant fire areas and the sensitivity results provided in the licensee's response to NRC RAI 8 was comprehensive and performed in sufficient detail to provide confidence that the licensee's use of IPEEE insights demonstrate that the impact of the ESW pump(s) AOT change in risk is acceptable and bounded by the internal events analysis.

Seismic, and Other External Risk Evaluation The IPEEE-Seismic program, integrated with the USI A-46 effort, resulted in several plant improvements and design modifications. The seismic PRA (SPRA) quantification in the IPEEE concluded that no severe accident vulnerabilities exist at Surry from a potential seismic event.

In its memorandum dated March 7, 2000 (ADAMS at Accession No. ML17142A410), the NRC concluded that on the basis of the IPEEE review, the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities. The IPEEE reported a seismic CDF of 8.0E-06/year.

The change in risk from the extended AOT is only affected if the seismic event occurs while the system is in the extended AOT, which the licensee estimates to be 15 days per year, or 0.04 years. The licensee clarified that in about 75 percent of the seismic PRA sequences, the SW supply to the Emergency Switchgear Room and Main Control Room chillers are either consequently failed by the seismic event or are not credited as support systems. Therefore the current seismic CDF that might credit the ESW pumps is equal to or less than 8E-08/year, and the LERF would generally be smaller. These results indicate that the seismic risk associated with the proposed change is very small.

The licensee also considered other external hazards consistent with NUREG/CR-2300 and NUREG/CR-4389. In its application dated July 14, 2016, the licensee stated, in part, that seven events were identified as needing more detailed evaluation following an initial screening. In its application, the licensee stated:

The non-seismic external events of interest, except for aircraft impacts, pipeline accidents and external flooding, were screened out based on the UFSAR information and the results reached by NUREG/CR-4550. The bounding analysis performed for the effects of aircraft impacts and pipeline accidents were based on the methods used by NUREG/CR-4550. The results of these two analyses indicate that the frequency of the events occurring is small. The actual risk from these hazards to the safe operation of the plant would be less than the screening value, because most safety-related equipment is inside Class I structures and is designed to withstand the loads imposed by the external event.

The bounding analysis for external flooding considered the worst case occurrence of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in 1 square mile probable maximum precipitation (PMP). The consequences of this occurrence were mitigated by implementation of a procedural revision and modification of turbine building.

Roof parapets to reduce roof top accumulation during intense precipitation.

Therefore, it can be concluded that non-seismic external events do not pose a significant risk to the safe operation of Surry Power Station.

Section 3.1 of RG 1.200, Revision 2 states that missing hazard groups may be evaluated using bounding arguments to cover the risk contributions not addressed by the model. In its application, the licensee summarized how it performed the screening analysis referencing acceptable NRC methods. The NRC staff concludes that the licensee's use of NRG-reviewed methods and models to evaluate the risk impact of the proposed change is sufficient to conclude that the evaluation is adequate to support the proposed change.

Tier 1: PRA Capability and Insights The licensee's evaluation addresses the NRC staff's three-tiered approach described in RG 1.177, Revision 1. The analysis evaluated the risk for one ESW pump out of service for durations in excess of the current TS 3.14.B limits. The first-tier evaluates the impact of the proposed change on plant operational risk. The Tier 1 review involves two aspects:

( 1) evaluation of the technical adequacy of the Surry PRA model and its application to the proposed change, and (2) evaluation of the PRA results and insights based on the licensee's proposed change.

The ESW pumps are explicitly modeled in the average maintenance model for Surry. The licensee analysis evaluated the ICCDP and the ICLERP for one ESW pump out of service for the new requested 14-day AOT. For the analysis, the licensee assumed that annual ESW

pump unavailability will increase by a factor of 2. The NRC Reactor Oversight Process Mitigating Systems Performance Indicator (MSPI) ESW pump unavailability data from the period of February 2013 through January 2016 was used as unavailability input.

The licensee's analysis evaluated the ICCDP and the ICLERP for the one ESW pump unavailable for a period of 14 days, which the licensee calculated to be:

U1 ICCDP U2 ICCDP U1 ICLERP U2 ICLERP Sin le 14-da TS 3.14.B Entr : 1.32E-11 1.99E-11 3.41 E-17 3.24E-16 Increasing these ICCDP and ICLERP values by a factor of two still results in a very small risk increase for CDF and LERF (i.e., less than 1E-6/year and 1E-07/year) that is within the acceptance criteria for RG 1.174, Revision 2 respectively. The NRC staff concludes that the proposed risk increase is very small; therefore, an additional evaluation of the total risk is unnecessary because the very small risk values are within the acceptable criteria for RG 1.174, and increases do not depend on the total risk.

Tier 2: Avoidance of Risk Significant Plant Configurations The licensee performed a detailed review of the PRA importance measures. The licensee stated that the detailed review did not identify any risk-significant maintenance configurations when one ESW pump is unavailable. The licensee concluded that the evaluation did not identify any configurations that would require Tier 2 enhancements in accordance with RG 1.177, Revision 1 (i.e., procedure revisions, and compensatory actions).

The staff concludes that the licensee's evaluation appropriately assessed the contribution to plant risk while the equipment covered by the proposed AOT change is out-of-service. The staff also concludes that the licensee's assessment for identifying that no risk significant configurations exist while the equipment covered by the proposed AOT change is out-of-service is consistent with the risk results supporting the conclusion of very limited risk impact from the requested change. Therefore, the staff concludes the licensee's evaluation is acceptable.

Tier 3: Risk-Informed Plant Configuration Control and Management The licensee stated that the 10 CFR 50.65(a)(4) program at Surry performs PRA analyses of planned maintenance configurations in advance. The licensee further stated that configurations that approach or exceed the NUMARC 93-01 risk limits are identified and either avoided or addressed by risk management actions. In addition, the configuration analysis and risk management processes are proceduralized in accordance with the requirements of 10 CFR 50.64(a)(4). As described below in key principle 5, the staff concludes that the licensee's program for compliance with 10 CFR 50.65(a)(4) ensures that the risk impact for out-of-service equipment is appropriately assessed and managed and is consistent with the guidance in RG 1.177, Revision 1.

3.2.2 Key Principle 5: Monitor the Impact of the Proposed Change RG 1.174, Revision 2 and RG 1.177, Revision 1 establish the need for an implementation and monitoring program to ensure that extensions to TS AOTs do not degrade operational safety over time and that no adverse degradation occurs due to unanticipated degradation or common cause mechanisms. An implementation and monitoring program is intended to ensure that the impact of the proposed TS change continues to reflect the reliability and availability of SSCs

impacted by the change. RG 1.174, Revision 2 states that monitoring performed in conformance with 10 CFR 50.65 can be used when the monitoring performed is sufficient for the SSCs affected by the risk-informed application.

The licensee stated that Surry's 10 CFR 50.65(a)(4) compliance program requires analysis and management of configuration risks in advance of planned maintenance configurations. The licensee also stated that the ESW pumps are included in the 10 CFR 50.65(a)(4) scope, and their removal from service will be monitored, analyzed, and managed. The staff concludes that the proposed change satisfies Key Principle 5 because the monitoring of the affected systems is accomplished with the 10 CFR 50.65 and, therefore, is acceptable.

3.3 Technical Evaluation Conclusion The NRC staff has reviewed the traditional engineering aspects of the licensee's evaluation related to the proposed changes to TS 3.14. Based on the results of the evaluation of traditional engineering considerations, the staff concludes that the proposed increase in the AOT to restore one inoperable ESW pump to operable status is consistent with current regulations, defense-in-depth attributes, and maintenance of adequate safety margins.

The NRC staff also concludes that the risk impact of the licensee's request to allow one ESW pump inoperable for up to 14 days, as estimated by ICCDP, ICLERP, ~CDF and ~LERF, is consistent with the acceptance guidelines specified in RG 1.177, Revision 1, RG 1.174, Revision 2, and the staff guidance outlined in Sections 19.1 and 16.1 of NUREG-0800. The licensee's methodology for assessing the risk impact is accomplished using PRA models of sufficient scope and technical adequacy based on a review of the model consistent with the guidance of RG 1.200, Revision 2. For external hazards that do not have PRA models, the licensee used bounding analyses. The NRC staff concludes that the licensee has followed the three-tiered approach and performance monitoring programs outlined in RG 1.177, Revision 1.

Therefore, the proposed changes satisfy both the traditional engineering and risk impact considerations delineated in RG 1.177 and are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the NRC staff notified the Virginia State official of the proposed issuance of the amendments on July 6, 2017. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and/or change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (81 FR 73443). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Adrienne Driver Steven Jones Date: July 28, 2017

D. Stoddard

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING THE EXTENSION OF THE EMERGENCY SERVICE WATER PUMP ALLOWED OUTAGE TIME, SURRY POWER STATION, UNITS 1 AND 2 (CAC NOS. MF8145 AND MF8146)

DATED JULY 28, 2017.

DISTRIBUTION:

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NAME KCotton KGoldstein SRosenburq JWhitman DATE 07/26/17 07/21/17 06/02/17 07/27/17 OFFICE NRR/DSS/SBPB/BC* OGC- NLO NRR/LPL2-1/BC NRR/LPL2-1/PM NAME RDenniq RNoorwood MMarkley (AKlett for) KCotton DATE 03/02/17 07/17/2017 07/27/17 07/28/17 OFFICIAL RECORD COPY